make changes from @paulromano's 3rd review

- syntax fixes and adjustments
- change name of FluxDepletionOperator to IndependentOperator
- flux_operator.py -> independent_operator.py
- new class, MicroXS, for creating (for now) one-group microscopic
  cross section DataFrames. This class takes the functionality that was
  previously in static functions in IndependentOperator
- Associated changes to the test suite and online docs

docs/source/usersguide/depletion.rst

@@ -97,7 +97,7 @@ Energy Deposition
 -----------------
 
 The default energy deposition mode, ``"fission-q"``, instructs the
-:class:`openmc.deplete.Operator` to normalize reaction rates using the product
+:class:`~openmc.deplete.Operator` to normalize reaction rates using the product
 of fission reaction rates and fission Q values taken from the depletion chain.
 This approach does not consider indirect contributions to energy deposition,
 such as neutron heating and energy from secondary photons. In doing this, the
@@ -113,7 +113,7 @@ should be, including indirect components. Some examples are provided below::
     fission_q = {"U235": 202e+6}  # energy in eV
 
     # create a Model object
-    model  = openmc.Model(geometry, settings)
+    model = openmc.Model(geometry, settings)
 
     # create a modified chain and write it to a new file
     chain = openmc.deplete.Chain.from_xml("chain.xml", fission_q)
@@ -133,7 +133,7 @@ to normalize reaction rates instead of using the fission reaction rates with::
         normalization_mode="energy-deposition")
 
 These modified heating libraries can be generated by running the latest version
-of :meth:`openmc.data.IncidentNeutron.from_njoy`, and will eventually be bundled
+of :meth:`openmc.data.IncidentNeutron.from_njoy()`, and will eventually be bundled
 into the distributed libraries.
 
 Local Spectra and Repeated Materials
@@ -188,31 +188,57 @@ Transport-independent depletion
    possible and likely in the near future.
 
 OpenMC supports running depletion calculations independent of the OpenMC
-transport solver using the :class:`~openmc.deplete.FluxDepletionOperator` class.
+transport solver using the :class:`~openmc.deplete.IndependentOperator` class.
 This class supports both constant-flux (``source-rate`` normalization) and 
 constant-power depletion (``fission-q`` normalization).
 
 .. important::
 
-   Make sure you set the correct parameter in the :class:`openmc.abc.Integrator` class. Use the ``source_rates`` parameter when ``normalization_mode == source-rate``, and use ``power`` or ``power_density`` when ``normalization_mode == fission-q``.
+   Make sure you set the correct parameter in the :class:`openmc.abc.Integrator`
+   class. Use the ``source_rates`` parameter when
+   ``normalization_mode == source-rate``, and use ``power`` or ``power_density``
+   when ``normalization_mode == fission-q``.
 
 .. warning::
 
-   The accuracy of results when using ``fission-q`` is entirely dependent on your depletion chain. Make sure it has sufficient data to resolve the dynamics of your particular scenario. 
+   The accuracy of results when using ``fission-q`` is entirely dependent on
+   your depletion chain. Make sure it has sufficient data to resolve the
+   dynamics of your particular scenario. 
 
-This class has two ways to initialize it: the default constructor accepts an 
-:class:`openmc.Materials` object and one-group microscopic 
-cross sections as a :class:`pandas.DataFrame`, while the ``from_nuclides``
-method accepts a volume and dictionary of nuclide concentrations in place of
-the :class:`openmc.Materials` object in addition to the other parameters.
-The class includes helper functions to construct the dataframe from a csv file
-or from data arrays::
+:class:`~openmc.deplete.IndependentOperator` class uses one-group microscopic cross sections to calculate reaction
+rates. Users can generate one-group microscopic cross sections using the 
+:class:`~openmc.deplete.MicroXS` class::
+
+    import openmc
+    from openmc.deplete import MicroXS
+
+    model = openmc.Model.from_xml()
+
+    micro_xs = MicroXS.from_model(model, model.materials[0])
+
+    micro_xs.to_csv(micro_xs_path)
+
+:class:`~openmc.deplete.MicroXS` also includes functions to read in cross
+section data directly from a ``.csv`` file or from data arrays::
+
+    micro_xs = MicroXS.from_csv(micro_xs_path)
+
+    nuclides = ['U234', 'U235', 'U238']
+    reactions = ['fission', '(n,gamma)']
+    data = np.array([[0.1, 0.2],
+                     [0.3, 0.4],
+                     [0.01, 0.5]])
+    micro_xs = MicroXS.from_array(nuclides, reactions, data)
+
+:class:`~openmc.deplete.IndependentOperator` has two ways to initialize it:
+the default constructor accepts an :class:`openmc.Materials` object and
+one-group microscopic cross sections as a :class:`~openmc.deplete.MicroXS`
+object, while the :meth:`~openmc.deplete.IndependentOperator.from_nuclides` 
+method accepts a volume and dictionary of nuclide concentrations in place of the
+:class:`openmc.Materials` object in addition to the other parameters::
 
-    ...
     # load in the microscopic cross sections
-    micro_xs = FluxDepletionOperator.create_micro_xs_from_csv(micro_xs_path)
-    flux = 1.16e15
-    op = FluxDepletionOperator(materials, micro_xs, chain_file)
+    op = IndependentOperator(materials, micro_xs, chain_file)
 
     # alternate construtor
     nuclides = {'U234': 8.92e18,
@@ -222,8 +248,8 @@ or from data arrays::
                 'O16': 4.64e22,
                 'O17': 1.76e19}
     volume = 0.5 
-    op = FluxDepletionOperator.from_nuclides(volume, nuclides, 'atom/cm3',
-                                             micro_xs, flux, chain_file)
+    op = IndependentOperator.from_nuclides(volume, nuclides, micro_xs,
+                                           chain_file, nuc_units='atom/cm3')
 
 A user can then define an integrator class as they would for a coupled
 transport-depletion calculation and follow the same steps from there.